ML17333A325

Criticality Analysis of DC Cook Nuclear Plant New Fuel Storage Vault W/Credit for Integral Fuel Burnable Absorbers, for Nov 1995

ML17333A325
Person / Time
Site:	Cook
Issue date:	11/30/1995
From:	Newmyer W, Savage C, Srinilta S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17333A324	List: ML17333A323 ML17333A325 ML17334B574
References
CDB-95-175, NUDOCS 9603010082
Download: ML17333A325 (27)

Text

CDB-95-175 Criticality Analysis of the Donald C. Cook Nuclear Plant New Fuel Storage Vault with Credit for Integral Fuel Burnable Absorbers November 1995 W. D. Newmyer S. Srinilta Prepared +COD '.

D. ewm er Critica ity Services Team Leader Verified:

S. Srinilta Core Design B Approved: < 7P.

C. R. Savage, Mana er Core Design B Westinghouse Commerical Nuclear Fuel Division ji 96030i0082 960226 g 050003i5i PDR ADQCK P PDR II

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List of Tables Table 1. Fuel P arameters Employed in the Criticality Analysis..........,.........................8 Table 2. Summary of Lattice Parameters for Strawbridge and Barry 101 Criticals ..........9 Table 3. B&W Core Loadings and Compositions Studied ................................10 Table 4. Donald C Cook Fresh Fuel Rack IFBA Requirement ......,..................................11 Donald C Cook Fresh Fuel Storage Racks

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List of Figures Figure 1.

~ Donald C Cook Nuclear Plant Fresh Fuel Rack Radial Layout................... 12 Figure 2. Donald C Cook Nuclear Plant Fresh Fuel Rack Axial Layout......, ....... 13 Figure 3. Donald C Cook Nuclear Plant Fresh Fuel Rack IFBA Requirement...................14 Donald C Cook Fresh Fuel Storage Racks ln

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1.0 Introduction This report presents the results of a criticality analysis of the American Electric Power Donald C. Cook Nuclear Plant New Fuel Storage Vault (NFSV) with credit for Integral Fuel Burnable Absorbers (IFBA).

The NFSV rack design considered herein is an existing array of Donald C. Cook Nuclear Plant NFSV unpoisoned racks previously qualified for storage of Westinghouse 15x15 STD and OFA',

and 17x17 STD and OFA fuel assemblies with maximum enrichments up to 4.55 w/o +5U.

In this analysis, credit for IFBA will be used to allow fuel enrichments up to 5.0 w/o + U.

The Donald C. Cook NFSV rack analysis is based on maintaining K ff < 0.95 under full water density conditions and 5 0.98 under low water density (optimum moderation) conditions.

1.1 Design Description The Donald C. Cook fresh fuel rack storage rack radial layout is depicted in Figure 1 on page 11 and the axial layout is shown in Figure 2 on page 12, with nominal dimensions provided on each figure.

The fuel parameters relevant to this analysis are given in Table 1 on page 8. With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the various types of Westinghouse 15x15 and 17x17 fuel (V5, VSH, V+, and P+) are beneficial in terms of extending burnup capability and improving fuel reliability, but do not contribute to any meaningful increase in the basic assembly reactivity. Therefore, future fuel assembly upgrades do not require a criticality analysis ifthe fuel parameters specified in Table 1 continue to remain bounding.

1.2 Design Criteria Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which

,limits fuel assembly interaction. This is done by fixing the minimum separation between fuel assemblies.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, K ff, of the NFSV when flooded with full density water will be less than 0.95 as recommended by ANSI 57.3-1983 and NRC guidance< >. Furthermore, the effective neutron multiplication factor, K ff, of the NFSV under low water density (optimum moderation) conditions will be less than 0.98 as recommended by NUREG-0800.

Introduction

2.0 Analytical Methods 2.1 Reactivity Equivalencing for IFBA Credit Storage of fuel assemblies with higher initial enrichments than 4.55 w/o U is achievable by means of the concept of reactivity equivalencing. Reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods. A series of reactivity calculations is performed to generate a set of enrichment-IFBA ordered pairs which all yield an equivalent K,ffwhen the fuel is stored in the Donald C. Cook Nuclear Plant fresh fuel racks. The data points on the reactivity equivalence curve are generated with the transport theory computer code, PHOENIX-P<3>. PHOENIX-P is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code which utilizes a 42 energy group nuclear data library.

PHOENIX-P has been used to demonstrate its predictive capability in a series of comparisons against direct experimental data from critical experiments. These comparisons provide a good assessment of the code's ability to predict key physics parameters over a wide range of lattice variations. Reactivity comparisons are accomplished by comparing appropriate predictions to Strawbridge and Barry's 101 criticals~"~ and the Babcock and Wilcox (B&W) cores XI-1,2,7,8,9 and cores XIV-1,6 spatial criticals< ' ~.

The range of lattice parameters of the 101 criticals are given in Table 2 on page 8. The resulting mean PHOENIX-P k ff for all 101 criticals is 1.00222 with a standard deviation of 0.00809. This shows PHOENIX-P results to be in excellent agreement with experimental data for all dependent parameters, with no significant bias or trends as a function of lattice parameters. For the specific set of lattice parameters used in this report, the standard deviation of PHOENIX-P K ff results would be much smaller thus allowing PHOENIX-P to be used for reactivity equivalencing calculations.

The core loadings and compositions studied in the seven B&W core spatial criticals are given in Table 3 on page 9. The resulting mean K ff of the seven critical experiments was 1.00151 with a standard deviation of 0.00098. The overall mean core K ff for this set of critical experiments with

. diverse lattice configurations is in very good agreement with expected experimental values. The overall standard deviation indicates excellent stability of the PHOENIX-P library and methodology.

Uncertainties associated with the IFBA dependent reactivities computed with PHOENIX-P are accounted for in the development of the individual reactivity equivalence limits. An uncertainty of approximately 10% of the total number of IFBA rods is accounted for in the development of the IFBA requirements. Additional information concerning the specific uncertainties included in the Donald C. Cook IFBA credit limit is provided in Section 3.2 of this report.

Analytical Methods

3.0 Criticality Analysis of Fresh Fuel Racks This section describes the analytical techniques and models employed to perform the criticality analysis and reactivity equivalencing evaluations for storage of mesh fuel in the Donald C. Cook Nuclear Plant New Fuel Storage Vault (NFSV).

Section 3.1 summarizes previous reactivity calculations performed for the fresh fuel storage racks with enrichments up to 4.55 w/o +5U and Section 3.2 describes the analysis which allows for storage of assemblies with enrichments above 4.55 w/o U+ and up to 5.00 w/o U by taking credit for Integral Fuel Burnable Absorbers (IFBAs).

3.1 Reactivity Calculations Previous calculations( ) have been performed for the Donald C. Cook Nuclear Plant New Fuel Storage Vault (NFSV) which show that fuel assemblies with enrichments up to 4.55 w/o U can be safely stored in the NFSV. The maximum 95/95 K ffdetermined for full water density flooding is 0.9495 and the maximum 95/95 K ff determined for optimum moderation flooding is 0.8974.

Based on these previously calculated K ff values, the acceptance criteria are met for both full and optimum water density flooding of the Donald C. Cook Nuclear Plant fresh fuel storage racks.

3.2 IFBA Credit Reactivity Equivalencing Storage of fuel assemblies with enrichments greater than '4.55 w/o +5U in the fresh fuel storage racks is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of Integral Fuel Burnable Absorbers (IFBAs). IFBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the UO2 fuel pellet. As a result, the neutron absorbing material is a non-removable or integral part of the fuel assembly once it is manufactured.

Two analytical techniques are used to establish the criticality criteria for the storage of IFBA fuel in the fresh fuel storage racks. The first method uses reactivity equivalencing to establish the poison material loading required to meet the criticality limits. The poison material considered in this analysis is a zirconium diboride (ZrB2) coating manufactured by Westinghouse. The second method uses the fuel assembly infinite multiplication factor to establish a reference reactivity. The reference reactivity point is compared to the fuel assembly peak reactivity to determine its acceptability for storage in the fresh fuel racks.

3.2.1 IFBA Requirement Determination A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent K,ff when the fuel is stored in the fresh fuel storage racks. The following assumptions were used for the IFBA rod assemblies in the PHOENIX-P models:

Criticality Analysis of Fresh Fuel Racks

1. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 17x17 OFA design (see Table 1 on page 7 for fuel parameters). For this calculation, the Westinghouse 17xl7 OFA assembly is the most reactive of the fuel assembly types considered.

2. The fuel assembly is modeled at its most reactive point in life.

3. The fuel pellets are modeled assuming 96% theoretical density and no dishing fraction.

4. No credit is taken for any natural enrichment or reduced enrichment axial blankets.

5. No credit is taken for any +4U or +~U in the fuel, nor is any credit taken for the buildup of fission product poison material.

6. No credit is taken for any spacer grids or spacer sleeves.

7. The IFBA absorber material is a zirconium diboride (ZrB~) coating on the fuel pellet. Each IFBA rod has a nominal poison material loading of 1.50 milligrams B per inch (1.0X IFBA loading), which is the minimum standard loading offered by Westinghouse for 17x17 OFA fuel assemblies.

8. The IFBA B loading is reduced by 5% to conservatively account for manufacturing tolerances and then by an additional 19% to conservatively model a minimum poison length of 116.9 inches, Calculations show that it is conservative to model IFBA in two dimensional geometry by smearing the IFBA poison length over the entire fuel stack.

9. The moderator is pure water (no boron) at a temperature of 68'F with a density of 1.0 gm/cm .

10. The array is infinite in lateral (x and y) and axial (vertical) extent. This precludes any neutron leakage from the array.

Figure 3 on page 13 shows the constant K,g contour generated for the fresh fuel storage racks.

Note the endpoint at 0 IFBA rods where the enrichment is 4.55 w/o and at 32(1.0X) IFBA rods where the enrichment is 5.00 w/o. The interpretation of the endpoint data is as follows: the reactivity of the fuel rack array when filled with fuel assemblies enriched to 5.00 w/o +~U with each containing 32(1.0X) IFBA rods is equivalent to the reactivity of the rack when filled with fuel assemblies enriched to 4.55 w/o and containing no IFBA rods. The data in Figure 3 on

'page 13 is also provided on Table 4 on page 10 for both 1.0X and 2.0X IFBA rods.

It is important to recognize that the curve in Figure3 on page13 is based on reactivity equivalence calculations for the specific enrichment and IFBA combinations in actual rack geometry (and not just on'imple comparisons of individual fuel assembly infinite multiplication factors). In this way, the environment of the storage rack and its influence on assembly reactivity is implicitlyconsidered.

The IFBA requirements of Figure 3 on page 13 were developed based on the standard IFBA patterns used by Westinghouse. However, since the worth of individual IFBA rods can change depending on position within the assembly (due to local variations in thermal flux), 'studies were performed to evaluate this effect and a conservative reactivity margin was included in the Criticality Analysis of Fresh Fuel Racks

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development of the IFBA requirement to account for this effect. This assures that the IFBA requirement remains valid at intermediate enrichments where standard IFBA patterns may not be available and provides margin for future changes in current standard IFBA patterns. In addition, to conservatively account for calculational uncertainties, the IFBA requirements of Figure 3 on page 13 also include a conservatism of approximately 10% on the total number of IFBA rods at the 5.00 w/o end (i,e., about 3 extra IFBA rods for a 5.00 w/o fuel assembly).

Additional IFBA credit calculations were performed to examine the reactivity efFects of higher IFBA linear B loadings (1.5X and 2.0X). These calculations confirm that assembly reactivity remains constant provided the net B material per assembly is preserved. Therefore, with higher IFBA B loadings, the required number of IFBA rods per assembly can be reduced by the ratio of the higher loading to the nominal 1.0X loading. For example, using 2.0X IFBA in 5.00 w/o fuel assemblies allows a reduction in the IFBA rod requirement from 32 IFBA rods per assembly to 16 IFBA rods per assembly (32 divided by the ratio 2.0X/1.0X).

3.2.2 Infinite Multiplication Factor The infinite multiplication factor, K, is used as a reference criticality reactivity point, and offers an alternative method for determining the acceptability of fuel assembly storage in the fresh fuel storage racks. The reference K is determined for a fresh 4.55 w/o fuel assembly.

The fuel assembly K calculations are performed using the PHOENIX-P computer code . The following assumptions were used to develop the infinite multiplication factor model:

1. The Westinghouse 17x17 OFA fuel assembly was analyzed (see Table 1 on page 8 for fuel parameters). The fuel assembly is modeled at its most reactive point in life and no credit is taken for any burnable absorbers in the assembly.

2. Allfuel rods contain uranium dioxide at an enrichment of 4.55 w/o U over the entire length of each rod.

3. The fuel array model is based on a unit assembly configuration (infinite in the lateral and axial extent) in Donald C. Cook reactor geometry (no rack).

4. The moderator is pure water (no boron) at a temperature of 68' with a density of 1.0 gm/cm~.

Calculation of the infinite multiplication factor for the Westinghouse 17x17 OFA fuel assembly in the Donald C. Cook core geometry resulted in a reference K of 1.4857. This includes a 1% AK reactivity bias to conservatively account for calculational uncertainties. This bias is consistent with the standard conservatism included in the Donald C. Cook core design r'efueling shutdown margin calculations.

Criticality Analysis of Fresh Fuel Racks

4.0 Summary of Criticality Results For the storage of fuel assemblies in the Donald C. Cook Nuclear Plant New Fuel Storage Vault, the acceptance criteria for criticality requires the effective neutron multiplication factor, K ~, to be less than or equal to 0.95, including uncertainties, under full water density flooded conditions and less than or equal to 0.98 under optimum moderation conditions.

This report and previous evaluationst I shows that the acceptance criteria for criticalhy is met for the Donald C. Co'ok Fresh Fuel Storage Racks for the stora e of Westinghouse 15x15 and 17x17 fuel assemblies with enrichments no greater than 4.55 w/o 5U with no IFBA and fuel assemblies with enrichments up to 5.0 w/o +5U with the minimum IFBA requirement of Section 3.2.

The analytical methods employed herein conform with ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7 Fuel Handling System; ANSI 57.3-1983, "Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants", ANSI N16.9-1975, "Validation of Calculational Methods for Nuclear Criticality Safety"; and the NRC Standard Review Plan, Section 9.1.1, "New Fuel Storage".

Summary of Criticality Results

Table 1. Fuel Parameters Employed in the Criticality Analysis Westinghouse Parameter Westinghouse Westingho use 15 15 STD 17x17 OFA 17x17 STD Number of Fuel Rods 204 264 264 per Assembly Rod Zirc-4 Clad O.D. (inch) OA22 0.360 0.374 Clad Thickness (inch) 0.0243 0.0225 0.0225 Fuel Pellet O.D.(inch) 0.3659 0.3088 0.3225 Fuel Pellet Density 96 96 96

(% of Theoretical)

Fuel Pellet Dishing Factor (%) 0.0 0.0 0.0 Rod Pitch (inch) 0.563 0.496 0.496 Number of Zirc-4 Guide Tubes 20 24 24 Guide Tube O.D. (inch) 0.533 0.474 0.482 Guide Tube Thickness (inch) 0.017 0.016 0.016 Number of Instrument Tubes Instrument Tube O.D. (inch) 0.533 0.474 0.484 Instrument Tube Thickness 0.017 0.016 0.016 (inch).

Donald C. Cook Fresh Fuel Storage Racks

Table 2. Summary of Lattice Parameters for Strawbridge and Barry 101 Criticals Lattice Parameter Range Enrichment (a/o) 1.04 to 4.07 Boron Concentration (ppm) 0 to 3392 Water to Uranium Ratio 1.0 to 11.96 Pellet Diameter (cm) OA4 to 2.35 Lattice Pitch (cm) 0.95 to 4.95 Clad Material none, aluminum, and stainless steel Lattice Type square and hexagonal Fuel Density (g/cm ) 7.5 to 18.9 Donald C. Cook Fresh Fuel Storage Racks

Table 3. BSW Core Loadings and Compositions Studied Number of Soluble Number of 'ater-Filled Number of Boron Core Loading Fuel Rods Positions Pyrex Rods (ppm) 4961 1511 4808 153 1334 4808 81 72 1031 4808 794 4808 779 XIV 4736 225 1289 4736 201 24 1179 Donald C. Cook Fresh Fuel Storage Racks

Table 4. Donald C. Cook Fresh Fuel Rack IFBA Requirement 1.0X (1.50 mg- B/in) 1.5X (2.25 mg- B/in) 2.0X (3.00 mg- B/in)

Enrichment IFBA Rods IFBA Rods IFBA Rods (w/o)

In Assembly In Assembly In Assembly 4.55 4.66 4.77 16 12 4.88 24 18 12 5.00 32 24 16 Donald C. Cook Fresh Fuel Storage Racks 10

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Bibliography

1. Nuclear Regulatory Comimssion, Letter to AllPower Reactor Licensees from B. K. Grimes, OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978.

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2. Melehan, J. B., Yankee Core Evaluation Program Final Report, WCAP-3017-6094, January 1971.

3. Nguyen, T. Q. et. al., Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11597-A, June 1988.

4. Strawbridge, L.E. and Barry, R.F., "Criticality Calculations for Uniform Water-Moderated Lattices," Nucl. Sci. Eng. 23, pp. 58-73 (1965).

5. Baldwin, M.N., and Stern, M.E., "Physics Verification Program Part III, Task 4: Summary Report," BAW-3647-20(March 1971),

6. Baldwin, M.N., "Physics Verification Program Part III, Task II: Quarterly Technical Report January-March 1974," BAW-3647-30(July 1974),

7. Baldwin, M.N., "Physics Verification Program Part III, Task II: Quarterly Technical Report l

July-September 974," BAW-3647-31(February 1975).

8. Boyd, WA., et. al., Criticality Analysis of the Donald C Cook Nuclear Plant fuel Racks, November 1989.

Bibliography 14

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